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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Analysis of fission-product energy release after thermal-neutron fission of {sup 235}U and {sup 239}Pu by scale-6.1.3 with recently available fission product yield data - 14419

The fission-product energy release after the reactor shutdown has been an important issue from the perspective of the nuclear reactor safety in a very short time. The energy release from the decay of fission products can be estimated using a summation calculation method implemented in various burnup and depletion programs. The method generally depends on the nuclear structure data such as the decay and fission product yield data. In this study, the decay heat measurements after thermal-neutron fission of {sup 235}U and {sup 239}Pu have been evaluated by the ORIGEN-S with the decay and fission product yield libraries included in the SCALE-6.1.3 code package. In order to verify the importance of the fission product yield data when estimating the decay heats, new fission product yields libraries for ORIGEN-S have been generated based on the recently available fission product yield data such as ENDF/B-VII.1, JEFF-3.1.1, JENDL/FPY-2011, and JENDL-4.0. The new libraries were applied to the decay heat calculations, and the results were compared with those of the reference calculation based on the yield library that belongs to the SCALE-6.1.3. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Rapid activation product separations from fission products and soil matrixes

We report that a rapid method for the separation and qualitative analysis of several neutron activation products ( 198 Au, 192 Ir, 72 Ga, 51 Cr, 191/195m/197 Pt, 54 Mn, 57 Co, and 59 Fe) from fission products and soil matrixes has been developed. Analytes were isolated within 20 h using ion exchange chromatography. After separation, the activation products were characterized by γ-spectroscopy and inductively coupled plasma-optical emission spectroscopy. Validation experiments demonstrated versatility of the method, showing that the activation products could be isolated from fresh fission products and other contaminants associated with complex soil matrixes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Angular Correlations of Gamma Rays from Fission Product Solution

Fission products from photon-induced fission are routinely used at Idaho National Laboratory as part of nuclear forensics activities. Typically, the focus is on fission products with relatively large fission yields (Ba-140, Mo-99, etc.). The nuclear data associated with the well-known isotopes have small uncertainties, and many have well-known angular correlations between emitted gamma-rays, if applicable. However, for isotopes with smaller fission yields or less intense gamma-ray emission intensities, the nuclear decay schemes and gamma-ray emission intensities may be less well-known. The Germanium Rotational Measurements for Angular Correlation (GeRMAC) system was designed to use two high-purity germanium (HPGe) detectors in a coincidence counting set-up but allows one of the HPGe detectors to rotate at a constant distance from a radioactive source, changing the relative angle between the two HPGe detectors. Angular correlations can then be determined from the relative photopeak areas between two gamma-rays emitted as part of the same nuclear decay. By recording the HPGe spectra in list mode, the angular correlations of gamma-rays emitted from complex sources, consisting of multiple isotopes, can be measured simultaneously and the individual isotope emissions can be isolated. Using the GeRMAC system, a multi-isotope fission product source, produced with photon-induced fission of U-238, was measured using a single rotational arc (180 degrees) to obtain the angular correlation between gamma-rays emitted by the large number of fission products contained in the solution with an angular accuracy of 1 degree. In particular, La-140 (38 gamma rays) and Nd-147 (26 gamma rays) are of interest to the nuclear forensics community.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2020

Measurement of distribution coefficients on TODGA resin and development of a group separation for mixed activation and fission product samples

Batch equilibrium experiments were performed to determine the distribution coefficients for Sc, Ir, and Ta on N,N,N’,N’tetraoctyl-1,5-diglycolamide (TODGA) resin in various acid matrices. Distribution coefficients were determined in hydrochloric acid (Sc, Ir), nitric acid (Sc, Ta), and a mixture of nitric acid and 0.2M hydrofluoric acid (Sc, Ir, Ta). The distribution coefficients determined through these experiments were then utilized to create a group separation method for activation products, fission products, and environmental constituents.

Louie, Christina A.↗
DOE OSTI2015

New Routes for the Production of Fission Products and Their Separation from the Source Material for the Development of Realistic Nuclear Forensic Debris

In order to fully characterize the NIF neutron spectrum, SAND-II-SNL software was requested/received from the Radiation Safety Information Computational Center. The software is designed to determine the neutron energy spectrum through analysis of experimental activation data. However, given that the source code was developed in Sparcstation 10, it is not compatible with current version of FORTRAN. Accounts have been established through the Lawrence Livermore National Laboratory’s High Performance Computing in order to access different compiles for FORTRAN (e.g. pgf77, pgf90). Additionally, several of the subroutines included in the SAND-II-SNL package have required debugging efforts to allow for proper compiling of the code.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Fission Product Yields from 232Th, 238U, and 235U Using 14 MeV Neutrons

Neutron-induced fission yield studies using deuterium-tritium fusion-produced 14 MeV neutrons have not yet directly measured fission yields from fission products with half-lives on the order of seconds (far from the line of nuclear stability). Fundamental data of this nature are important for improving and validating the current models of the nuclear fission process. Cyclic neutron activation analysis (CNAA) was performed on three actinide targets–thorium-oxide, depleted uranium metal, and highly enriched uranium metal–at the University of Michigan's Neutron Science Laboratory (UM-NSL) using a pneumatic system and Thermo-Scientific D711 accelerator-based fusion neutron generator. This was done to measure the fission yields of short-lived fission products and to examine the differences between the delayed fission product signatures of the three actinides. The measured data were compared against previously published results for 89Kr, -90, and -92 and 138Xe, -139, and -140. The average percent deviation of the measured values from the Evaluated Nuclear Data Files VII.1 (ENDF/B-VII.1) for thorium, depleted-uranium, and highly-enriched uranium were -10.2%, 4.5%, and -12.9%, respectively. In addition to the measurements of the six known fission products, 23 new fission yield measurements from 84As to 146La are presented.

Pierson, Bruce D.↗
DOE OSTI2020

Analysis of fission product distribution and composition in the TRISO layers of AGR-2 fuel

The Advanced Gas Reactor Fuel Qualification and Development Program's second irradiation experiment focuses on the performance of fuel particles produced in an industrial-scale coating system. Ongoing post-irradiation examination efforts include exploration of tristructural-isotropic (TRISO)-coated particle fuel in the as-irradiated state and subjected to elevated temperatures to simulate accident conditions (safety testing). Scanning electron microscopy (SEM) analysis with energy dispersive x-ray spectroscopy (EDS) was used to show the distribution of fission products within the TRISO layers. The SEM analysis showed a variation of fission product distribution within the TRISO layers based on irradiation temperature, burnup, individual particle fission product inventory retention, safety testing history and kernel composition. Finally, a key observation is that particles which likely experienced higher temperature showed a greater diversity and distribution of fission products and actinides in the IPyC and SiC layers, implying that temperature is a dominant variable influencing the transport of radionuclides in TRISO particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Fission Product Measurement and Off-Gas Scrubbing for Molten Salt Reactors

The Advanced Reactor Technology (ART) program within the US Department of Energy is developing understanding and technology to help facilitate molten salt reactor commercialization. The off-gas system for MSRs plays an important role in fission product confinement. The strategy used in MSRs is different from that used in pressurized water reactors, where the off-gas mainly mitigates severe accident conditions. Although the details vary depending on the reactor type, the off-gas system of MSRs must reduce radiological emissions to the environment to levels comparable with those mandated for non-MSRs. Systems with molten fuel salt will accumulate material in the headspace: volatile fission products, entrained salt particles, graphite fines, and material that has low solubility in the salt. Operators of thermal fluoride salt reactors with encapsulated fuel will also be concerned about tritium generation and migration. The fission and activation products include noble gases, their daughters, tritium, and aerosolized salt constituents. Their concentrations will generally increase with reactor burnup, depending on the various strategies used for mitigation. Additionally, tritium capture and removal has been reviewed by several groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Energy Dependence of Fission Product Yields from {sup 235}U, {sup 238}U and {sup 239}Pu for Incident Neutron Energies Between 0.5 and 14.8 MeV

Fission Product Yields (FPY) have historically been one of the most observable features of the fission process. They are known to have strong variations that are dependent on the fissioning species, the excitation energy, and the angular momentum of the compound system. However, consistent and systematic studies of the variation of these FPY with energy have proved challenging. This is caused primarily by the nature of the experiments that have traditionally relied on radiochemical procedures to isolate specific fission products. Although radiochemical procedures exist that can isolate all products, each element presents specific challenges and introduces varying degrees of systematic errors that can make inter-comparison of FPY uncertain. Although of high importance in fields such as nuclear forensics and Stockpile Stewardship, accurate information about the energy dependence of neutron induced FPY are sparse, due primarily to the lack of suitable monoenergetic neutron sources. There is a clear need for improved data, and to address this issue, a collaboration was formed between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL) and the Triangle Universities Nuclear Laboratory (TUNL) to measure the energy dependence of FPY for {sup 235}U, {sup 238}U and {sup 239}Pu. The measurements have been performed at TUNL, using a 10 MV Tandem Van de Graaff accelerator to produce monoenergetic neutrons at energies between 0.6 MeV to 14.8 MeV through a variety of reactions. The measurements have utilized a dual-fission chamber, with thin (10-100 μg/cm2) reference foils of similar material to a thick (100-400 mg) activation target held in the center between the chambers. This method allows for the accurate determination of the number of fissions that occurred in the thick target without requiring knowledge of the fission cross section or neutron fluence on target. Following activation, the thick target was removed from the dual-fission chamber and gamma-ray counted using shielded HPGe detectors for a period of 1-2 months to determine the yield of various fission products. To the extent possible all irradiation and counting procedures were kept the same to minimize sources of systematic errors. FPY have been determined at incident neutron energies of 0.6, 1.4, 2.4, 3.5, 4.6, 5.5, 8.9 and 14.8 MeV.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Waste form evaluation for RECl 3 and REO x fission products separated from used electrochemical salt

The work presented here is based off the concept that the rare earth chloride (RECl3) fission products mixture within the used electrorefiner (ER) salt can be selectively removed as RECl3 (not yet demonstrated) or precipitated out as REOCl through oxygen sparging (has been demonstrated). This paper presents data showing the feasibility of immobilizing a mixture of RECl3’s at 10 mass% into a TeO2-PbO glass and it shows that this same mixture of RECl3’s can be oxidized to REOCl at 300°C and then to REOx by 1200°C. When the REOx mixture is heated at temperatures >1200°C, the ratios of REOx’s change. The mixture of REOx was then immobilized in a LABS glass at a high loading of 60 mass%. Both the TeO2-PbO glass and LABS glass systems show good chemical durability. The advantages and disadvantages of tellurite and LABS glasses are compared.

Riley, Brian J. (ORCID:0000000277456730)↗
DOE OSTI2015

Cement As a Waste Form for Nuclear Fission Products: The Case of 90 Sr and Its Daughters [Cement As a Container for Nuclear Fission Products: The Case of 90 Sr and Its Daughters]

One of the main challenges faced by the nuclear industry is the long-term confinement of nuclear waste. Because it is inexpensive and easy to manufacture, cement is the material of choice to store large volumes of radioactive materials, in particular the low-level medium-lived fission products. It is therefore of utmost importance to assess the chemical and structural stability of cement containing radioactive species. Here, we use ab initio calculations based on density functional theory (DFT) to study the effects of 90 Sr insertion and decay in C–S–H (calcium-silicate-hydrate) in order to test the ability of cement to trap and hold this radioactive fission product and to investigate the consequences of its β-decay on the cement paste structure. We show that 90 Sr is stable when it substitutes the Ca 2+ ions in C–S–H, and so is its daughter nucleus 90 Y after β-decay. Interestingly, 90 Zr, daughter of 90 Y and final product in the decay sequence, is found to be unstable compared to the bulk phase of the element at zero K but stable when compared to the solvated ion in water. Furthermore, cement appears as a suitable waste form for 90 Sr storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Characterization of U-Zr fuel with alloying additive Sb for immobilizing fission product lanthanides

Fission product lanthanides can migrate to the fuel periphery and enhance fuel-cladding chemical interaction (FCCI). Adding Sb into the fuel alloy to bind lanthanides and thus to prevent lanthanides from migration has been proposed. The U-Zr fuel alloys were fabricated with compositions of U-10Zr-4.1Sb, U-10Zr-4.1Sb-4Ce, U-10Zr-2.07Sb, and U-10Zr-2.07Sb-2Ce wt.% by arc melting, and analyzed under scanning electron microscope (SEM) for the microstructure and elemental compositions. In the Ce-free alloy, Sb was found in Sb-Zr precipitates; while in the Ce-contained alloy, Sb was found in the form of intermetallic compounds Ce 4 Sb 3 and Ce 2 Sb without any Sb-Zr precipitates. In conclusion, antimony has the preference to bind Ce over Zr, indicating Sb can effectively prevent Ce interaction with Fe to the extent. Cerium is bound to Sb and no ternary intermetallic Ce-Sb-Zr compound was found.

36 MATERIALS SCIENCE↗
NASA NTRS1967

Computer program FPIP-REV calculates fission product inventory for U-235 fission

Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

Brown, W. S.↗
DOE OSTI2020

Zinc-in-titania waste form for immobilizing lanthanide fission products from electrochemical reprocessing

This paper provides microstructural, microchemical, and bulk X-ray diffraction information for a zinc-in-titania waste form for immobilizing rare-earth oxide (REOx) fission products recovered through precipitation from electrochemical salt wastes. This waste form can be used to immobilize REOx at loadings as high as 40 mass% in a chemically durable matrix consisting of glass-bonded minerals including Zn2TiO4 as well as RE-containing phases of REOx, RE2Ti2O7, and REPO4 (monazite). Chemical durability data is also provided, and comparisons are drawn between this waste form and one made using a similar process with half the REOx loading.

rare earth, rare earth waste form, electrochemical↗
DOE OSTI2024

Energy dependence of fission product yields in the second-chance fission region

An extensive dataset of cumulative fission product yields has been generated under a joint collaboration between Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and the Triangle Universities Nuclear Laboratory (TUNL). The energy dependence of the cumulative yield for a select number of high-yield fission products has been measured using quasimonoenergetic neutrons with energies between 5.5 and 11.0 MeV. This is in addition to previously published data covering 0.5–4.5 and 14.8 MeV. The absolute number of fissions was determined during the irradiation period using dual-fission ionization chambers, and the fission products were measured postactivation by whole target γ-ray spectroscopy. Here this paper presents the absolute cumulative fission product yields as a function of incident neutron energy from the neutron-induced fission of 235 U, 238 U, and 239 Pu isotopes at four incident energies in the second-chance fission region and compares them with existing literature values. Corrections relevant to this collaboration's previously published data are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Fission Product and Actinide Distribution in the SiC Layer of Fissioned Triso Fuel Particles

Tri-structural Isotropic (TRISO) encapsulated fuel particles are used in next generation high-temperature gas-cooled nuclear reactors. The integrity of the fuel over a wide range of temperatures is integral to the safety of an operating nuclear reactor. It has been observed that fission products are released at high temperatures that simulate hypothesized accident scenarios. The particles under study were from the AGR-1 fuel development and qualification program. They consist of an enriched UCO fuel kernel surrounded by a dense carbon buffer layer (100 {mu}m), an inner pyrolytic carbon (IPyC) layer (40 {mu}m), a silicon carbide (SiC) layer (35 {mu}m), and an outer pyrolytic carbon (OPyC) layer (40 {mu}m). The SiC layer acts as the main diffusion barrier to fission products, but has been demonstrated to fail due to release of Ag through intact particles and local corrosion from metallic fission products such as Pd. In this work, we use transmission electron microscopy and diffraction techniques to study the structure and dispersion of release products within the SiC of TRISO particles as a function of annealing temperature used to simulate margin temperatures for in-reactor accident scenarios. In order to improve our understanding of fission product retention as a function of temperature, we have the unique opportunity to examine irradiated, fissioned TRISO particles to study post-irradiation effects on these materials. This data will contribute to current models used to predict the behavior of fission products in the SiC layer of TRISO fuels.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Characterization of a platform for the gas transport, collection, and identification of fission products in the high-intensity laser environment

Recent progress in the production of laser accelerated high flux proton beams opens new possibilities for the study of fission in unique environments. To this end, we are currently pursuing the development of a platform for the swift gas transport, collection, and identification of fission products. Fission is induced in targets fixed inside a sealed chamber through which a carrier gas flows, transporting fission products to a carbon filter for collection and online spectroscopy. This has been recently demonstrated using target normal sheath accelerated protons at the PHELIX laser facility. There were large discrepancies between measured rates and those expected based on established fission yields and measured beam parameters. These discrepancies prompted a large number of tests at the Idaho Accelerator Center (IAC), where fission in uranium is induced by bremsstrahlung photons generated by 21 MeV electrons from a linear electron accelerator. Here, the results are compared to a series of models that account for the slowing of energetic fission fragments in the carrier gas and the fluid dynamics of the gas flow that transports fission products to the collection filter. Characterization of the apparatus reveals a few mechanisms that together account for a portion of the previously observed discrepancies at the PHELIX laser facility. However, additional research is necessary before high-accuracy experiments can be performed with the apparatus. Work performed under the auspices of the U.S. Department of Energy by LLNL under contract DE-AC52-07NA27344.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗